Licensing the ACR in the USA - A Status Report
- 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)
The ACR-700 (ACR) is an evolutionary reactor design, which incorporates the inherent safety features of the Candu products, as well as the successful operating experience of the current Candu 6 reactors. The improvements to the ACR from Candu 6 result in significant reductions in capital and operating costs as well as enhanced safety. AECL Technologies (AECLT, a wholly owned US subsidiary of Atomic Energy of Canada Limited) is the proponent for the ongoing pre-application review of the ACR with the US Nuclear Regulatory Commission (NRC) in the United States. This pre-application review will be completed shortly and will support an application to the USNRC for Standard Design Certification (SDC). AECL Technologies' overall objective for the pre-application review of the ACR is to obtain an understanding of the scope, cost, and the schedule to obtain a Standard Design Certification for the ACR. The pre-application review will address licensing issues associated with the Candu reactor technologies in ACR that depart from the light water reactor, pressure-vessel based regulatory framework in the USA. Therefore, during the course of the ACR pre-application review, major USNRC issues with the ACR design will be identified early and the scope of the work required to address these concerns, along with associated completion schedules, will be formulated and ultimately agreed upon with the USNRC. AECLT has been informed by the NRC staff that the results of their pre-application review will be documented in a Safety Assessment Report (SAR), which will state whether there are any major impediments to licensing the ACR in the United States. In particular, the SAR should provide confirmation of the licensing criteria applicable to the ACR, provide an assessment of the completeness of AECL's Research and Development (R and D) programs that exist or are planned in support of the ACR, provide an assessment of the suitability for purpose of the computer codes used in the safety analysis of the ACR, and provide estimates of the cost and schedule for the remaining scope of the USNRC's efforts on the identified focus topics in preparation for a Design Certification review. This paper will discuss the status of the pre-application review for ACR. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21160683
- Resource Relation:
- Conference: ICAPP'04: 2004 international congress on advances in nuclear power plants, Pittsburgh, PA (United States), 13-17 Jun 2004; Other Information: Country of input: France; Related Information: In: Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04, 2338 pages.
- Country of Publication:
- United States
- Language:
- English
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